Biweekly Notice

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving

No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from November 28, 1994, through December 9, 1994.

The last biweekly notice was published on December 7, 1994.

Notice Of Consideration Of Issuance Of Amendments To Facility

Operating Licenses, Proposed No Significant Hazards Consideration

Determination, And Opportunity For A Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, and should cite the publication date and page

number of this Federal Register notice. Written comments may also be

delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,

Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.

Copies of written comments received may be examined at the NRC Public

Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.

The filing of requests for a hearing and petitions for leave to

intervene is discussed below.

By January 13, 1995, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, Attention: Docketing and Services

Branch, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

date. Where petitions are filed during the last 10 days of the notice

period, it is requested that the petitioner promptly so inform the

Commission by a toll-free telephone call to Western Union at 1-(800)

248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator

should be given Datagram Identification Number N1023 and the following

message addressed to (Project Director): petitioner's name and

telephone number, date petition was mailed, plant name, and publication

date and page number of this Federal Register notice. A copy of the

petition should also be sent to the Office of the General Counsel, U.S.

Nuclear Regulatory Commission, Washington, DC 20555, and to the

attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN

50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit

Nos. 1, 2, and 3, Maricopa County, Arizona

Date of amendment requests: November 2, 1994

Description of amendment requests: The proposed amendment would

delete the Condenser Vacuum Exhaust release point reference on Figure

5.1-3 and combine it with the Plant Vent Exhaust release point on the

revised Figure 5.1-3. In addition to the figure change, Bases Section

3/4.3.3.6 is amended to note the deletion of radiation monitor RU-142

and the relocation of RU-144 and RU-146 from Table 3.3-13 (previously

deleted) to the Offsite Dose Calculation Manual (ODCM).

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis about the issue of no significant hazards

consideration, which is presented below:

Standard 1--Does the proposed change involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Combining the condenser vacuum and the plant vent exhausts has

no affect

[sic] on the operation of the radiation monitoring system or its

intended functions. Routing of the condenser vacuum exhaust to the

plant vent exhaust is in the same area as the old system and does

not affect accident initiation or consequences. The change has no

affect

[sic] on the operation of the plant. The radiation monitors

affected by this change do not provide engineered safety features or

protection system actuation signals. Therefore, the change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

Standard 2 -- Does the proposed change create the possibility of

a new or different kind of accident from any accident previously

evaluated?

The overall system is designed to assist the operators in

evaluating and controlling the radiological consequences of normal

plant operations, anticipated operational occurrences, and

postulated accidents. The change does not affect the way the system

is operated. Therefore, the change does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

Standard 3 -- Does the proposed change involve a significant

reduction in a margin of safety?

Combining of the condenser vacuum and plant vent exhaust into a

single release path does not involve a significant reduction in a

margin of safety. The change involves the removal of one high range

monitor in the condenser vent, however, its function is provided by

the high range monitor in the plant vent. The ranges of the monitors

are the same. The existing plant effluent radiation monitors will

serve to monitor both the plant and condenser air removal system

effluent. The normal range monitors have the ability to adequately

detect radiation over five decades and these monitors will stay in

place and they have the ability to perform the anticipated radiation

release detection. Therefore, the proposed change does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

that review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 12

East McDowell Road, Phoenix, Arizona 85004

Basis for proposed no significant hazards consideration

determination: I11Attorney for licensees: Nancy C. Loftin, Esq.,

Corporate Secretary and Counsel, Arizona Public Service Company, P.O.

Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999

NRC Project Director: Theodore R. Quay

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-

324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick

County,North Carolina

Date of amendments request: November 16, 1994 Description of

amendments request: The proposed Technical Specification (TS) change

would (1) revise TS 4.6.1.2 by removing the schedular requirements for

Type A overall integrated leakage rate tests to be performed at 40 plus

or minus 10 month intervals and replacing the acceptance criteria for

these Type A integrated leakage rate tests with a reference to the

containment integrated leakage testing requirements of Appendix J to 10

CFR Part 50, (2) delete TS 4.6.1.2.a through TS 4.6.1.2.c because they

are no longer needed, (3) revise TS 4.6.1.2.h to remove the prohibition

against applying TS 4.0.2 to the 40 plus or minus 10 month integrated

leakage rate test frequency, (4) delete Unit 1 one-time footnote *

located on TS page 3/4 6-3A and on Table 4.6.1.2-1 listed on TS page 3/

4 6-3B since the exception provision has expired, (5) delete Unit 1

one-time footnote ** located on TS page 3/4 6-3A since the exception

has expired, (6) delete Unit 2 footnote * located on TS page 3/4 6-3

because the exception constitutes an approved exemption.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

. The proposed amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed amendments remove the detailed technical and

schedular information pertaining to primary containment integrated

leakage rate testing from the Technical Specifications and

references the corresponding requirements that are located in the

Appendix J to 10 CFR Part 50. As such, the proposed amendments are

an administrative change since the actual requirements for the

performance of primary containment integrated leakage rate testing

are not being changed. No safety-related equipment, safety function,

or plant operations will be altered as a result of the proposed

amendments. The change does not affect the design, materials, or

construction standards of the primary containment nor the test

methods, test acceptance criteria, or testing frequencies applicable

to primary containment integrated leakage rate testing. Based on the

above, the proposed license amendments do not create a significant

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed amendments would not create the possibility of a

new or different kind of accident from any accident previously

evaluated. As stated above, no safety-related equipment, safety

function, or plant operations will be altered as a result of the

proposed change. The proposed amendments do not change the primary

containment design or the test methods, test acceptance criteria, or

testing frequencies for primary containment integrated leakage rate

testing. As such, the proposed license amendments cannot create the

possibility of a new or different kind of accident from any accident

previously evaluated.

3. The proposed amendments do not involve a significant

reduction in the margin of safety. The proposed amendments do not

involve any changes to the test methods, acceptance criteria, or

testing frequency for primary containment integrated leakage rate

testing. Thus, the proposed amendments will not affect the ability

of the primary containment to perform its intended safety function

and no margins of safety, as defined by the plant's accident

analyses, are impacted. Primary containment integrated leakage rate

testing will continue to be performed in accordance with the

regulatory requirements of Appendix J to 10 CFR Part 50. Based on

the above reasoning, the proposed license amendments do not involve

a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Attorney for licensee: R. E. Jones, General Counsel, Carolina Power

& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602

NRC Project Director: William H. Bateman

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: September 28, 1994

Description of amendment request: Amendments will update the ``Loss

of Power'' functional unit of the Engineered Safety Features Actuation

System (ESFAS) Instrumentation tables within the Technical

Specifications for McGuire Nuclear Station.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

As required by 10 CFR 50.91, this analysis is provided

concerning whether the requested amendments involve significant

hazards considerations, as defined by 10 CFR 50.92. Standards for

determination that an amendment request involves no significant

hazards considerations are if operation of the facility in

accordance with the requested amendment would not: 1) Involve a

significant increase in the probability or consequences of an

accident previously evaluated; or 2) Create the possibility of a new

or different kind of accident from any accident previously

evaluated; or 3) Involve a significant reduction in a margin of

safety.

The requested amendments update the existing one-level

undervoltage protection to be exclusively for loss of voltage, and

add a second level of undervoltage protection to be exclusively for

degraded voltage.

In 48 FR 14870, the Commission has set forth examples of

amendments that are considered not likely to involve significant

hazards considerations. Example vi describes a change which either

may result in some increase to the probability or consequences of a

previously-analyzed accident or may reduce in some way a safety

margin, but where the results of the change are clearly within all

acceptable criteria with respect to the system or component

specified in the Standard Review Plan. The requested amendments are

similar to example vi in that they result in some increase to the

probability of a previously-analyzed accident, the Loss of Offsite

Power accident, but where the changes are clearly based on the

recommendations of Branch Technical Position PSB-1.

Criterion 1

The requested amendments will not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The requested amendments will involve some increase in

the probability of an accident previously evaluated. Automatic

separation from offsite power (a LOOP accident) will be more

probable because the voltage setpoints for the new relaying will be

higher than the settings for the existing relaying. The closer relay

settings are to 100% bus voltage, the more frequently actual bus

voltage can be expected to occur at or below the setpoint. The

occurrence of a LOOP presents a challenge to safety systems. More

probable (e.g., more frequent) LOOPs increase the frequency of

safety system challenges, which increases the probability of

malfunction of equipment important to safety. However, offsetting

this probability increase is a probability decrease due to the

protection of safety equipment from degraded voltage conditions,

given by the added protective relaying. The EPC system is required

to provide power for equipment used for accident mitigation and safe

shutdown. The ability of the EPC system to perform its required

safety functions will not be degraded by the implementation of this

TS change. No common failure modes are created between redundant EPC

system power trains. Therefore, the consequences of an accident or

malfunction of equipment important to safety evaluated in the SAR

are not increased.

Criterion 2

The requested amendments will not create the possibility of a

new or different kind of accident from any accident previously

evaluated. No new failure modes are created by the implementation of

this TS change. No accidents previously considered incredible are

made credible. The added protective relaying is expected to be as

reliable as the existing relaying. The added equipment is QA

Condition 1, and qualifications of equipment enclosures have been

maintained. Thus, the possibility of an accident or malfunction of

equipment of a different type than evaluated in the SAR will not be

created.

Criterion 3

The requested amendments will not involve a significant

reduction in a margin of safety. The setpoints for the existing Loss

of Power protective relays are lowered by this TS change. The new

setpoints have been evaluated and will not prevent the protective

relaying from performing its required safety function. The fission

product barriers (RCS pressure boundary, containment, fuel pellets,

and cladding) are not degraded. No assumptions made in any accident

analysis are affected by the implementation of this TS change,

except as previously discussed for probability of a Loss of Offsite

Power. Therefore, the margin of safety as defined in the basis for

any Technical Specification is not decreased.

Based on the preceding analyses, Duke Power concludes that the

requested amendments do not involve a significant hazards

consideration.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Atkins Library, University of

North Carolina, Charlotte (UNCC Station), North Carolina 28223

Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422

South Church Street, Charlotte, North Carolina 28242

NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-

412,Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,

Pennsylvania

Date of amendment request: August 31, 1994

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs), Section 6, Administrative

Controls, and includes line-item improvements suggested by Generic

Letter 93-07. The proposed changes include the following:

1. Elimination of the references to specific frequencies for

each of the Technical Specification required audits.

2. Elimination of the references to reviews and audits of the

Emergency Plan and Security Plan.

3. Separation of the Inservice Inspection (ISI) and Inservice

Testing (IST) Programs surveillance requirements and removal of the

requirement that relief requests be granted before they are

implemented for both IST and ISI.

4. Editorial changes which were necessitated by a

reorganization.

5. Elimination of the reference to Appendix A of 10 CFR Part 55.

6. Elimination of the requirement to perform an independent fire

protection and loss prevention program inspection annually.

7. Inclusion of the Offsite Dose Calculation Manual and Process

Control Program and associated implementing procedures into the list

of required audits.

8. Updates of the Beaver Valley Power Station (BVPS) Unit 2

License Conditions to reflect completion of activities.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The likelihood that an accident will occur is neither increased

or decreased by this proposed Technical Specification change which

only affects review and audit frequencies, removes redundancies in

the audit program, corrects editorial information, and updates the

Unit 2 license conditions. This Technical Specification change will

not impact the function or method of operation of plant equipment.

Thus, there is not a significant increase in the probability of a

previously analyzed accident due to this change. No systems,

equipment, or components are affected by the proposed change. Thus,

the consequences of a malfunction of equipment important to safety

previously evaluated in the Updated Final Safety Analysis Report are

not increased by this change.

The proposed change affects audit frequencies, types of audits

listed in the technical specifications, references for some

technical specification sections, the time frame for Inservice

Testing (IST) and Inservice Inspection (ISI) relief request

submittals, and editorial changes necessitated by an internal

reorganization. As such, the proposed change has no impact on

accident initiators or plant equipment, and therefore, does not

affect the probabilities or consequences of an accident.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed technical specification revisions do not involve

changes to the physical plant or operations. Since program audits,

organizational titles, and technical specification references do not

contribute to accident initiation, a change related to the areas

listed in the description section [***] cannot produce a new

accident scenario or produce a new type of equipment malfunction.

Therefore, this change does not alter any existing accident

scenarios. The proposed change does not affect equipment or its

operation.

Therefore, the proposed change does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

3. Does the change involve a significant reduction in a margin

of safety?

The proposed change concerns the conduct of audits, technical

specification references, ISI and IST relief request submittals,

completed License conditions, and organizational title changes and

does not directly affect plant equipment or operation. Safety limits

and limiting safety system settings are not affected by this

proposed change.

Therefore, use of the proposed Technical Specification would not

involve a significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment involves no significant hazards consideration.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Alquippa, Pennsylvania 15001.

Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,

Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,

Washington, DC 20037.

NRC Project Director: Walter R. Butler

Entergy Operations Inc., Docket No. 50-382, Waterford Steam

ElectricStation, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: August 11, 1994, as supplemented

December 2, 1994.

Description of amendment request: The requested change would remove

cycle-specific variables from the Waterford 3 Technical Specifications

(TSs) and control them under a new document called the Core Operating

Limits Report (COLR). All cycle-specific limits that are to be included

in the COLR must be calculated using NRC approved methodologies. The

proposed change is consistent with the TS line-item improvement

guidelines provided by the NRC in Generic Letter (GL) 88-16, ``Removal

of Cycle-Specific Parameter Limits From Technical Specifications,''

dated October 3, 1988.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Removing cycle-specific variables from the TS and placing

them into a COLR, is consistent with the NRC guidance provided in GL

88-16. These changes are administrative in nature and have no impact

on plant operation or accident analyses. The TS will continue to

require operation within the core operational limits for each cycle

reload calculated by the approved reload methodologies. If these

limits are violated, Technical Specifications will continue to

ensure that the appropriate actions are taken.

The cycle-specific evaluation demonstrates that changes in the

fuel cycle design and the corresponding COLR do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Listing the NRC approved methodologies in the COLR as opposed to

the TS Administrative Controls section is purely an administrative

change in contrast to NUREG 1432. The proposed change requires the

use of NRC approved methodologies. Listing the approved

methodologies in the TS provides the potential for an increased

licensee and NRC administrative burden without a commensurate

increase in safety or control.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed changes, to relocate the cycle-specific

variables from TS to the COLR, are administrative in nature. No

change in the design, configuration, or method of operation of the

plant is made by this amendment. The cycle-specific variables will

continue to be calculated using NRC approved methods. TS will

continue to require operation within the required core operating

limits and appropriate actions will be taken if the limits are

exceeded.

Listing the NRC approved methodologies in the COLR as opposed to

the TS Administrative Controls section is purely an administrative

change in contrast to NUREG 1432. The proposed change requires the

use of NRC approved methodologies. Listing the approved

methodologies in the TS provides the potential for an increased

licensee and NRC administrative burden without a commensurate

increase in safety or control.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The margin of safety presently provided is not affected by

removing cycle-specific core operating limits from TS. The core

limits contained in the COLR are obtained through analyses using NRC

approved methodologies. The TS still: (1) require that the core be

operated within these limits and (2) specify appropriate actions to

be taken if the limits are violated. The cycle-

specific COLR limits for future reload will also be developed

based on NRC-approved methodologies. In addition, each reload will

involve a 10CFR 50.59 safety review to assure that operation of the

unit within the cycle-specific limits will not involve a reduction

in a margin of safety.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

Safety and Significant Hazard Determination

Based on the above safety analysis, it is concluded that: (1)

the proposed change does not constitute a significant hazards

consideration as defined by 10CFR50.92; and (2) there is a

reasonable assurance that the health and safety of the public will

not be endangered by the proposed change; and (3) this action will

not result in a condition which significantly alters the impact of

the station on the environment as described in the NRC Final

Environmental Statement.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122

Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L

Street N.W., Washington, D.C. 20005-3502

NRC Project Director: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,

Appling County, Georgia

Date of amendment request: October 13, 1994

Description of amendment request: The proposed amendments would

revise the Hatch Technical Specifications (TS) as follows:

1. Lower the anticipated transient without scram-recirculation pump

trip (ATWS-RPT) setpoint by approximately 2 feet 2 inches to minimize

the potential for recirculation pump trips following reactor scrams.

2. Allow restarting the recirculation pump following an RPT when

the temperature differential between the coolant at the reactor bottom

head and the reactor steam dome cannot be obtained, provided certain

conditions are met.

The licensee believes the above changes will aid in preventing

thermal stratification and unnecessary thermal cycles resulting from

the rapid cooldown of the bottom head region and the reduction in

reactor pressure to atmospheric conditions.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

Proposed Change 1

Proposed Change 1 does not involve a significant hazards

consideration, because it does not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Lowering the ATWS-RPT trip will not increase the probability of

occurrence of any design basis accident or transient, since this

change does not physically affect any component of the reactor

coolant pressure boundary (RCPB). Therefore, the probability of a

LOCA event is not increased. Lowering the ATWS-RPT water level

setpoint does not increase the probability of an ATWS event, since

no component of the CRD system or the reactor protection system is

being physically altered by this change. Also, the operation of

these two systems is not affected.

Reducing the setpoint may require installation of new slave trip

units; however, this addition does not increase the probability of

occurrence of accidents or transients. The new trip units will be

functionally identical to other slave trip units already in use at

Plant Hatch and are within the design capabilities of ATTS. In

conclusion, no safety-related plant system or component is being

affected in a manner that would render it more susceptible to

failure.

Lowering the setpoint does not result in an increase of the

consequences of a previously evaluated accident. GE reviewed the

proposed reduction and determined the results of the ATWS event with

the lowered setpoint remain acceptable. An approved analytical

method (REDY) was used to evaluate a bounding ATWS event -- LOFW

[loss of feedwater]. The results indicate that reactor power with

the new ATWS-RPT setpoint remains stable, with no unacceptable power

spikes. Hot and cold reactor shutdowns can still be ultimately

attained.

The consequences of non-ATWS events are not increased. For LOCA

events, reducing the recirculation pump low water level trip

setpoint allows the recirculation pumps to run longer. The forced

circulation provided by the recirculation pumps keeps the fuel

cooler for a longer period of time. The ECCS-LOCA analysis assumes

the pump trip and coastdown early in the event. Therefore, lowering

the ATWS-RPT makes the ECCS-LOCA analysis more conservative and, as

a result, it does not need to change.

Based on the above discussion, Proposed Change 1 does not

constitute an increase in the probability or consequences of a

previously analyzed accident.

2. Create the possibility of a new or different kind of accident

from any previously evaluated.

Lowering the ATWS-RPT trip will not alter the design or

operation of any safety-related system. The change may require

adding new slave trip units to ATTS; however, the new trip units

will be functionally identical to the equipment already in use at

Plant Hatch. Furthermore, the addition of this slave trip unit is

within the design capabilities of ATTS.

Since no new operation modes, accident scenarios, or failure

modes are introduced, Proposed Change 1 does not create the

possibility of a new type of accident.

3. Involve a significant reduction in the margin of safety.

As stated previously, reducing the ATWS-RPT low water level

setpoint will not cause unacceptable results for ATWS events.

Specifically, the LOFW event is bounding for all the ATWS events. An

evaluation using approved analytical methods indicates that reducing

the ATWS-RPT setpoint will not result in power instabilities or

unacceptable power spikes, or prevent the mitigation of an ATWS

event. (Reference Enclosure 1 [of the licensee's submittal],

Proposed Change 1).

The ATWS-RPT aids in maintaining the level above the top of the

active fuel. The reduction of core flow reduces the neutron flux and

thermal power and, therefore, the rate of coolant boil-off. However,

the setpoint reduction does not significantly reduce the margin of

safety since a substantial margin remains to the top of the active

fuel.

For non-ATWS events, delaying the RPT will provide a slight

improvement in the current ECCS-LOCA analysis, thereby improving the

margin of safety.

The margin of safety for transients is not reduced because plant

transient (MCPR) analyses do not take credit for the ATWS-RPT trip.

Proposed Change 2

Proposed Change 2 does not involve a significant hazards

consideration, because it does not:

1. Involve a significant increase in the probability of

occurrence or the consequences of a previously analyzed accident.

Allowing a recirculation pump restart within 30 minutes of a

trip, when the temperature differential is unknown, will not

increase the probability of occurrence of a previously analyzed

accident because this change does not physically alter the RCPB.

Additionally, the proposed change does not alter the design or

function of any safety-related systems.

Furthermore, no recirculation system equipment is being changed

as a result of this amendment. The start circuitry and trip

circuitry remain[s] unaffected. Operation of the recirculation

system with the reactor at power is also unaffected. As a result,

the probability of the chapter 14 and 15 events dealing with the

recirculation system are not increased; i.e., trip of one or both

recirculation pumps, recirculation pump seizure, recirculation flow

controller failure, etc.

The purpose of the 145 deg.F temperature differential

requirement is to avoid thermal shock caused by hot water on the

cold CRD stub tubes during recirculation pump restart. If the

temperature differential is unable to be determined, restart within

30 minutes of the trip will not increase the probability or severity

of thermal fatigue on the stub tubes. As discussed in Enclosure 1,

Basis for Proposed Change 2, stratification will not develop within

a 30-minute period following pump trip, thus, the temperature

differential will not exceed 145 deg.F. Additional caveats are

provided to insure the required temperature differential is met.

These involve certain conditions of ECCS injection, feedwater

temperature, and drive flow.

General Electric verified that this provision for recirculation

pump restart will not affect any plant safety analysis, including

radiological analysis. Therefore, the consequences of previously

analyzed events are not increased.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The proposed recirculation pump restart provisions do not

introduce any new plant operating modes, accident scenarios, or

equipment failure modes. All other requirements for recirculation

pump restart; e.g., those addressing equipment protection and power

oscillations, will continue to apply.

3. Involve a significant reduction in the margin of safety.

The 145 deg.F differential temperature requirement is in place

to avoid thermal fatigue on the CRD stub tubes and the in-core

housing welds. Allowing the early restart with the listed caveats,

when temperature indication is not available, is acceptable because

the conditions for re-start insure that a stratified condition has

not yet developed. Thus, the cooler vessel structures at the vessel

bottom will not experience a severe thermal shock resulting from

exposure to hot water following the pump restart.

This change will actually aid in preventing the development of a

stratified condition, since the recirculation pumps will be

restarted before a stratified condition can develop, thereby helping

to maintain RCPB integrity. In the past, it has often been necessary

to depressurize the RPV [reactor pressure vessel] to atmospheric

pressure before the required temperature differential was met.

Proposed Change 2 should reduce the number of times depressurization

is required, thus avoiding unnecessary thermal cycles on the RPV.

Therefore, the margin of safety regarding the protection of RPV

components from severe thermal stresses, and the integrity of the

RCPB has not been reduced, and may actually increase.

The margin of safety in existing plant analyses is not reduced,

because none of the analyses are adversely affected as a result of

allowing the pump restart within 30 minutes of the RPT, as indicated

in GE's review of plant transient and accident analyses.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia 31513

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037

NRC Project Director: Herbert N. Berkow

Georgia Power Company, Oglethorpe Power Corporation, Municipal

Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.

50-424 and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1

and 2, Burke County, Georgia

Date of amendment request: October 3, 1994

Description of amendment request: This amendment would replace the

reactor coolant system heatup and cooldown limitations for VEGP Units 1

and 2, contained in Technical Specification figures 3.4-2a through 3.4-

3b, and the maximum allowable nominal power-operated relief valve

(PORV) setpoint for the cold overpressure protection system. These

changes are the results of new analyses that account for the

nonconservatisms identified in NRC Information Notice 93-58, the

results of reactor pressure vessel surveillance capsule examinations,

and recently issued ASME Code Case N-514.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Conformance of the proposed amendment with the standards for a

determination of no significant hazards, as defined in the three

factor test of 10 CFR 50.92, is shown in that the proposed

amendment:

1. Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The revised heatup and cooldown limits and PORV setpoints ensure

that the Appendix G pressure/temperature limits are not exceeded and

therefore, help ensure that RCS integrity is maintained. The changes

do not result in a condition where the design, material, and

construction standards of the RCS are altered. In addition, the

safety function of the COMS (cold over-pressure mitigation system),

which is related to accident mitigation, has not been degraded.

Therefore, the probability of an accident is not increased by the

PORV setpoint change.

The changes do not adversely affect the integrity of the RCS

such that its function in the control of radiological consequences

is affected. In addition, the changes do not affect any fission

barrier. The changes do not degrade or prevent the response of the

COMS or other safety-related system to accident scenarios, as

described in FSAR chapter 15. In addition, the changes do not alter

any assumption previously made in the radiological consequence

evaluations nor affect the mitigation of the radiological

consequences of an accident described in the FSAR. Therefore, the

consequences of an accident previously evaluated in the FSAR will

not be increased.

Thus, operation of VEGP Units 1 and 2 in accordance with the

proposed license amendment, does not involve a significant increase

in the probability or consequences of any accident previously

evaluated.

2. Does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

The changes do not cause the initiation of any accident nor

create any new credible limiting single failure for safety-related

systems and components. The changes do not result in any event

previously deemed incredible being made credible. As such, it does

not create the possibility of an accident different than any

evaluated in the FSAR.

The changes do not have any effect on the ability of the safety-

related systems to perform their intended safety functions. The

changes do not create failure modes that could adversely impact

safety-related equipment. Therefore, it will not create the

possibility of a malfunction of equipment important to safety

different than previously evaluated in the FSAR. Thus, the proposed

license amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. Does not involve a significant reduction in a margin of

safety.

The evaluation has shown that the PORV setpoints ensure that the

Appendix G pressure/temperature limits are not exceeded. The

analysis to support the proposed PORV setpoint change demonstrates

that the appropriate criteria, including that of ASME Code Case N-

514, are met for the postulated RCS pressures and temperatures. An

adequate margin of safety against vessel failure is assured, in

part, by the safety factors identified in Appendix G to Section III

of the ASME Boiler and Pressure Vessel Code, and [in] the basis for

ASME Code Case N-514 as well as [in the] added margin to prevent

lifting of the PORVs. The heatup and cooldown limits are designed to

prevent nonductile failure of the reactor vessel and take into

account the results of surveillance capsule Y on the reactor vessel

materials for VEGP Unit 1. The actuation of the safety-related

components and responses of the safety-related systems will remain

as modeled in the safety analyses. The changes will have no adverse

[effect] on the availability, operability, or performance of the

COMS. Therefore, the changes will not reduce the margin of safety,

as described in the bases to any Technical Specification.

Thus, [this] proposed license amendment does not involve a

significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia 30830.

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia 30308

NRC Project Director: Herbert N. Berkow

Gulf States Utilities Company, Cajun Electric Power Cooperative,

and Entergy Operations, Inc., Docket No. 50-458, River Bend

Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: January 14, 1994, as supplemented by

letter dated November 10, 1994.

Description of amendment request: The proposed amendment would

revise the technical specifications (TSs) by removing component lists

from the TSs in accordance with NRC Generic Letter (GL) 91-08 and by

removing the schedule for withdrawal of reactor vessel material

specimen capsules from the TSs in accordance with GL 91-01. This

proposed amendment was originally noticed in the Federal Register on

May 23, 1994, (59 FR 26675). The licensee's letter dated November 10,

1994, provides clarification of the wording in the proposed TSs and

does not change the proposed determination that the amendment request

involves no significant hazards consideration. However, the notice is

being repeated here.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change will not result in any hardware or operating

changes. The proposed change is based upon Generic Letters 91-01 and

91-08 and merely removes component lists, removes details relating

to the component lists, provides clarifying information supporting

the removal of the component listings, or removes details (which are

considered administrative) that are no longer applicable to the

Technical Specifications. The components listed in the affected

Technical Specifications are assumed in the mitigation of accident

and transient events. The removal of tabular component listings from

the Technical Specifications does not impact affected component

OPERABILITY requirements. Technical Specifications will continue to

require the components to be OPERABLE. Action statements and

surveillance requirements for the components will also remain in the

Technical Specifications. The tabular component lists are relocated

to the Technical Requirements Manual which will be in accordance

with the change control provisions specified in the Administrative

Controls Section of the Technical Specifications (Specification

6.5.2). Therefore, this change is administrative in nature and does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change does not necessitate a physical alteration

of the plant (no new or different type of equipment will be

installed) or changes to parameters governing normal plant

operation. The proposed change will not impose any different

requirements and adequate control of information will be maintained.

No new failure modes are introduced. Therefore, this proposed change

does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed change will not reduce a margin of safety because

it has no impact on any safety analysis assumption. The proposed

changes do not alter the scope of equipment currently required to be

OPERABLE or subject to surveillance testing, nor do the proposed

changes affect any instrument setpoints or equipment safety

functions. Therefore the change does not involve a significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, Louisiana 70803

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005

NRC Project Director: William D. Beckner

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,

Linn County, Iowa

Date of amendment request: November 10, 1994

Description of amendment request: The proposed amendment revises

the Duane Arnold Energy Center (DAEC) Technical Specification (TS)

Section 3.2.A to refer to the Offsite Dose Assessment Manual (ODAM) for

the setpoint of the Offgas Stack Radiation Monitor and makes the

``Applicable Operating Mode'' and the ``Action'' statements for these

instruments consistent with the required function. The Action statement

for the other instruments which initiate Secondary Containment

isolation is also revised to be consistent with the current practice

and with the function of those instruments. The Basis is also revised

to add further description of the function and requirements.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is provided below:

1) The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated because the instruments will still be required

to be operable to initiate an isolation at a setpoint which will

assure that the offsite dose limits are preserved, as designed, or

else administrative controls will be established for the venting of

primary containment. Through either means, offsite releases will be

maintained within the limits established in the ODAM. The change to

the applicable operating mode simply will require that the

instruments be operable when they are assumed to be operable in

previously analyzed accidents. The change to the required action

when the TS requirement cannot be met will assure that the flow path

from containment is isolated or that positive control is established

so that any offsite radioactive gaseous release is within the limits

analyzed in the ODAM.

2) The proposed amendment will not create the possibility of a

new or different kind of accident from any previously evaluated

because the affected instruments are inputs to the secondary

containment isolation and the revised specification will assure that

they are operable or adequately compensated when they are assumed to

perform their function. The instruments initiate a secondary

containment isolation in the event that high radiation levels are

detected in the monitored effluent.

3) The proposed amendment will not involve a significant

reduction in a margin of safety because the revised applicability

statement will assure that the instruments are operable when they

are required to perform their function. The proposed compensatory

action allows administrative control of the isolation valves when

the instruments are inoperable and it is necessary to continue

venting. This allowance recognizes that venting is a controlled

evolution and that operator action would be adequate to prevent

excessive releases in the event of high radioactivity in the offgas

piping. The revision to the setpoint will not affect system

operation, but will continue to assure that the gaseous effluents

released are within the limits specified in the ODAM.

In summary, the proposed changes do not change the probability

or consequences of an accident previously evaluated, do not create

the possibility for a new or different kind of accident and do not

involve a reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

thisreview, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cedar Rapids Public Library,

500 First Street, S.E., Cedar Rapids, Iowa 52401.

Attorney for licensee: Jack Newman, Kathleen H. Shea, Newman,

Bouknight & Edgar, PC, 1615 L Street, NW., Washington, DC 20036.

NRC Project Director: Leif J. Norrholm

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of amendment requests: October 7, 1994

Description of amendment requests: The proposed amendments would

remove the requirements for the Nuclear Safety and Design Review

Committee (NSDRC) to audit, and for the Plant Nuclear Safety Review

Committee (PNSRC) to review, the Emergency and Security plans and

implementing procedures. The composition of the PNSRC and the NSDRC

would also be revised to reflect organizational changes. Changes would

be made to the delegation of responsibility by the Site Vice President/

Plant Manager, and title corrections would be made on all pages

affected by the above changes.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:We [the licensee] have

evaluated the proposed T/S changes and have determined that the changes

should involve no significant hazards consideration. The proposed

amendment involves changes to the administrative controls section of

the T/Ss only. Because all changes reflect organizational/title changes

only or guidance from GL 93-07, they do not:

1) involve a significant increase in the probability or

consequence of an accident previously evaluated;

2) create the possibility of a new or different kind of accident

from any accident previously evaluated; or

3) involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske

MemorialLibrary, 500 Market Street, St. Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and Trowbridge, 2300 N Street, NW, Washington, DC 20037

NRC Project Director: John N. Hannon

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of amendment requests: November 16, 1994

Description of amendment requests: The proposed amendments would

allow core offload 100 hours after core subcriticality instead of the

168 hours currently required. Also included in this submittal are minor

typographical corrections to Figure 5.6-1, ``Normal Storage Pattern

(Mixed Three Zone), and Figure 5.6-2, ``Interim Storage Pattern

(Checkerboard).

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

We [the licensee] have evaluated the proposed T/S, editorial and

clarification changes and have determined that they do not represent

a significant hazards consideration based on the criteria

established in 10 CFR 50.92(c). Operation of Cook Nuclear Plant in

accordance with the proposed amendment will not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Although one of the proposed changes results in initiation of

core offload earlier after subcriticality than is currently allowed,

it does not increase the probability or consequences of an accident

previously evaluated. The bulk pool water temperatures, fuel rod

clad temperatures, and pool wall concrete temperatures will be

within acceptable limits as shown in Attachment 2 [of the November

16, 1994, submittal]. In addition, the subject change will not

result in an uncontrolled release of radiation to the environment

and will not initiate an accident. The remaining changes are

editorial in nature and have no [e]ffect on probability or

consequences of a postulated accident.

(2) Create the possibility of a new or different kind of

accident from an accident previously evaluated.

As previously stated, the earlier fuel movement change will not

result in bulk pool water, fuel rod clad, or concrete temperatures

which would initiate bulk pool boiling, challenge fuel rod integrity

or jeopardize the structural integrity of the pool. This change will

also have no impact on the criticality, structural, seismic, or

dropped assembly accident analysis previously performed and accepted

by the NRC. Consequently, the proposed T/S change does not create

the possibility of a new or different kind of accident from any

previously analyzed. The remaining changes have no [e]ffect on [the]

nature or probability of a postulated accident.

(3) Involve a significant reduction in a margin of safety.

The proposed change for earlier fuel movement will not result in

bulk pool water temperatures, fuel rod clad temperatures or concrete

temperatures which would initiate bulk pool boiling, challenge fuel

rod integrity or jeopardize the structural integrity of the pool.

This proposed change will not affect the results of any other

analysis associated with the spent fuel pool. It is, therefore,

concluded that this change poses no significant reduction in a

margin of safety. The remaining changes have no [e]ffect on the

nature or probability of a postulated accident.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and Trowbridge, 2300 N Street, NW, Washington, DC 20037

NRC Project Director: John N. Hannon

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,

Michigan

Date of amendment requests: November 18, 1994

Description of amendment requests: The license amendment requests

propose a change to Technical Specification (T/S) 4.0.5 for both units

to delete the wording ``except where specific written relief has been

granted by the Commission pursuant to 10 CFR 50, Section

50.55a(g)(6)(i).'' This change, which is consistent with guidance in

the November 1993 draft NUREG-1482, ``Guidelines for Inservice Testing

at Nuclear Power Plants,'' would allow the licensee to implement

certain 10 CFR 50.55a relief requests while the relief requests were

being reviewed by the NRC.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

We [the licensee] have evaluated the proposed T/S change and

have determined that the change involves no significant hazards

consideration. Operation of Cook Nuclear Plant in accordance with

the proposed amendment will not:

(1) Involve a significant increase in the probability or

consequences of an accident.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The inspections required under Section XI are intended to

show the operational readiness of the applicable components, and

exceptions to the Code are allowed. When taking relief from Code

requirements, alternate requirements are developed which provide a

high level of confidence that components will perform their intended

function.

The proposed change does not alter the Code requirements or

lessen our obligations under existing regulations. Its only effect

is to allow implementation of Code relief prior to obtaining NRC

written approval. The proposed T/S change is consistent with NUREG-

1431, and, as such, has been found to be acceptable by the NRC.

Therefore, we believe that implementation of this change will not

involve a significant increase in the probability or consequences of

a previously analyzed incident.

(2) Create the possibility of a new or different kind of

accident from any previously analyzed.

The proposed amendment does not create the possibility of a new

or different kind of accident from any previously evaluated. Typical

relief requests involve using alternative testing methods or

increasing the time interval between tests. Each proposed

alternative must assure that the component will perform its intended

function. The proposed change involves no physical changes to the

plant; therefore, we believe that implementation of this change will

not introduce a new of different kind of accident than previously

analyzed.

(3) Involve a significant reduction in a margin of safety.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The inspections required under Section XI are intended to

show the operational readiness of the applicable components, and

exceptions to the Code are allowed. When taking relief from Code

requirements, alternate requirements are developed which provide a

high level of confidence that components will perform their intended

function.

The proposed change does not altar the Code requirements or

lessen our obligations under existing regulations. Its only effect

is to allow implementation of Code relief prior to obtaining NRC

written approval. The proposed T/S change is consistent with NUREG-

1431, and, as such, has been found to be acceptable by the NRC.

Therefore, we believe that implementation of this change will not

result in a significant reduction of the margin of safety.

NRC staff has reviewed the licensee's analysis and, based on this

review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Maud Preston Palenske Memorial

Library, 500 Market Street, St. Joseph, Michigan 49085

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and Trowbridge, 2300 N Street, NW, Washington, DC 20037

NRC Project Director: John N. Hannon

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile

Point Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: November 14, 1994

Description of amendment request: The proposed license amendment

would revise Technical Specification 4.5.1.e.2.e) to reduce the leak

rate test pressure for the Automatic Depressurization System (ADS)

nitrogen receiving tanks from 385 psig to 365 psig. This pressure

reduction would be made to reduce potential degradation of the rupture

disk installed on each ADS nitrogen receiving tank during periodic leak

testing of the receiving tanks. Plant operating experience has shown

that leak rate testing at 385 psig occasionally results in inadvertent

failure of the rupture disks. Testing at the reduced pressure would be

consistent with the manufacturer's recommendations.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The ADS is required to effect or support the safe shutdown of

the reactor. This is accomplished by the blowdown of steam from a

group of seven designated ADS SRVs [safety/relief valves] to the

suppression pool. The proposed change to the test pressure does not

affect any accident precursors. Therefore, the proposed change

cannot increase the probability of an accident previously evaluated.

In the event the nitrogen gas supply from the nitrogen gas

storage tanks is lost, a minimum nitrogen pressure of 334 psig in

the ADS nitrogen receiver tanks assures a five-day supply of

nitrogen to the ADS accumulators. The proposed change to

Surveillance Requirement 4.5.1.e.2.e) would decrease the leak rate

test pressure of the ADS nitrogen receiver tanks from 385 psig to

365 psig. Since the proposed test pressure remains well above the

design minimum pressure of 334 psig, the surveillance test continues

to ensure that the actual leakage of the safety related ADS

accumulator pneumatic supply system is bounded by the leakage

assumptions contained in the system design. In addition, the

surveillance test continues to ensure that the ADS nitrogen receiver

tanks are capable of providing a 5-day supply of nitrogen to the ADS

accumulators. Therefore, the proposed change does not significantly

increase the consequences of a previously evaluated accident.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed leak rate test pressure of 365 psig for the ADS

nitrogen receiver tanks is above the minimum design pressure to

assure a 5-day supply of nitrogen is available to the ADS

accumulators if makeup from the high pressure nitrogen gas storage

tanks is lost. With the proposed change, the ADS will continue to

perform its safety function of effecting and supporting the safe

shutdown of the reactor. The nitrogen receiving tank test pressure

is not a precursor for any new or different accident and the change

does not affect the operation of the system in any way.

Accordingly, the proposed change does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The operation of the ADS SRVs, in conjunction with the LPCI [low

pressure coolant injection] mode of RHR [residual heat removal

system] and/or the LPCS [low pressure core spray] system, functions

as an alternative to the HPCS [high pressure core spray system] for

protection against fuel cladding damage upon a small break loss-of-

coolant accident. The blowdown of steam by these SRVs depressurizes

the reactor, allowing injection by the low-pressure coolant

injection sources. With the proposed change, the ADS will continue

to perform its intended safety function of effecting and supporting

the safe shutdown of the reactor as an alternate to the HPCS. The

proposed test pressure of 365 psig remains well above the minimum

acceptable pressure for the nitrogen receiver tanks of 334 psig.

Therefore, the change will not involve a significant reduction in

the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW., Washington, DC 20005-3502.

NRC Project Director: Michael J. Case

Northeast Nuclear Energy Company, et al., Docket No. 50-336,

Millstone Nuclear Power Station, Unit No. 2, New London County,

Connecticut

Date of amendment request: November 30, 1994

Description of amendment request: The proposed amendment would

modify the Technical Specifications by adding a footnote to Limiting

Conditions for Operation (LCOs) 3.8.1.1.b and 3.8.1.2.b which will

denote that 24,000 gallons of fuel oil is capable of supporting the

operation of one emergency diesel generator (EDG) for at least 4 days

and the other EDG for 1 hour with the EDGs loaded to the continuous

rated load of 2750 kW.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

...The proposed changes do not involve a significant hazards

consideration because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases

Section 3/4.8 will revise the Millstone Unit No. 2 design

requirements regarding the volume of EDG fuel oil which is required

to be stored onsite. The new rationale indicates that 24,000 gallons

of safety-related fuel oil would support the operation of one EDG

for at least four days with the other EDG running for at least one

hour. These run-times assume the EDGs are loaded to the continuous

rated loading of 2750 kW.

The proposed changes have no effect on EDG operation and

reliability. They provide additional operational flexibility,

because the EDG loading can be varied without the EDG minimum run-

time being altered. Also, an EDG run-time of at least four days

provides significant time to replenish fuel oil from onsite and

offsite sources even in the event of a hurricane or seismic event.

Based on the above, there is no effect on the probability or

consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The EDGs are required to operate in response to a loss of

offsite power. The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b

and Bases Section 3/4.8 do not change the manner in which the EDGs

respond to a design basis accident. Also, the proposed changes do

not introduce any new failure mechanisms. Therefore, the proposed

changes do not create the possibility of a new or different kind of

accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases

Section 3/4.8 have no effect on EDG operation and reliability. They

provide additional operational flexibility, because the EDG loading

can be varied without the EDG minimum run-time being altered.

An EDG run-time of at least four days provides significant time

to replenish EDG fuel oil via onsite or offsite sources even in the

event of a hurricane of seismic event. EPIP 4400 requires that the

need to order EDG fuel oil be evaluated within four hours of a loss

of offsite power event. Also, the high reliability of the electrical

grid and the high probability that offsite power would be restored

within 24 hours reduces the need to rely on extended EDG operation.

Millstone Unit No. 2 has more margin than is indicated by the

new design requirements. The EDG run-time will be significantly

greater than four days, because the electrical loading on the EDGs

will be less than the continuous rated loading, and electrical loads

will be shed through normal recovery actions following a design

basis accident.

Based on the above, the proposed changes do not involve a

significant reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resource Center,

Three Rivers Community-Technical College, Thames Valley Campus, 574 New

London Turnpike, Norwich, CT 06360.

Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,

Northeast Utilities Service Company, Post Office Box 270, Hartford, CT

06141-0270.

NRC Project Director: Phillip F. McKee

Northern States Power Company, Docket Nos. 50-282 and 50-306,

Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue

County, Minnesota

Date of amendment requests: July 11, 1994

Description of amendment requests: The proposed amendments would

change license condition 2.C.4 of each license to conform to the

standard fire protection license condition as stated in Generic Letter

(GL) 86-10, ``Implementation of Fire Protection Requirements.'' In

addition, the amendments would delete the fire protection program

elements from the Technical Specifications and incorporate, by

reference, the NRC-approved Fire Protection Program and major

commitments, including the fire hazards analysis, into the Updated

Safety Analysis Report. Guidance for these proposed changes is also

provided in GL 88-12, ``Removal of Fire Protection Requirements from

Technical Specifications.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed amendment[s] will not involve a significant

increase in the probability or consequences of [an] accident

previously evaluated.

The requested changes are administrative in nature in that they

move fire protection requirements from the Technical Specifications

to the Fire Protection Program and associated implementing

procedures following the guidance provided in GL 86-10 and GL 88-12.

The requested changes will not revise the requirements for fire

protection equipment operability, testing or inspections. The

amendment would give added responsibility to the Operations

Committee for review of the Fire Protection Program in accordance

with the guidance given in GL 86-10 and 88-12 including special

reporting requirements associated with limiting conditions for

operation for fire protection systems.

The proposed changes do not involve any change to the

configuration or method of operation of any plant equipment that is

used to mitigate the consequences of an accident, nor do they affect

any assumptions or conditions in any of the accident analyses. Since

the accident analyses remain bounding, their radiological

consequences are not adversely affected.

Therefore, the probability or consequences of an accident

previously evaluated are not affected.

(2) The proposed amendment[s] will not create the possibility of

a new or different kind of accident from any accident previously

analyzed.

The requested changes are administrative in nature in that they

move fire protection requirements from the Technical Specifications

to the Fire Protection Program and associated implementing

procedures following the guidance provided in GL 86-10 and GL 88-12.

The requested changes will not revise the requirements for fire

protection equipment operability, testing or inspections. The

amendment would give added responsibility to the Operations

Committee for review of the Fire Protection Program in accordance

with the guidance given in GL 86-10 and 88-12 including special

reporting requirements associated with limiting conditions for

operation for fire protection systems.

The proposed changes do not involve any change to the

configuration or method of operation of any plant equipment that is

used to mitigate the consequences of an accident.

Therefore, the possibility of a new or different kind of

accident from any accident previously evaluated would not be

created.

(3) The proposed amendment[s] will not involve a significant

reduction in the margin of safety.

The requested changes are administrative in nature in that they

move fire protection requirements from the Technical Specifications

to the Fire Protection Program and associated implementing

procedures following the guidance provided in GL 86-10 and GL 88-12.

The requested changes will not revise the requirements for fire

protection equipment operability, testing or inspections. The

amendment would give added responsibility to the Operations

Committee for review of the Fire Protection Program in accordance

with the guidance given in GL 86-10 and 88-12 including special

reporting requirements associated with limiting conditions for

operation for fire protection systems.

Therefore, a significant reduction in the margin of safety would

not be involved.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and

Trowbridge, 2300 N Street, NW, Washington, DC 20037

NRC Project Director: John N. Hannon

Omaha Public Power District, Docket No. 50-285, Fort Calhoun

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: November 11, 1994

Description of amendment request: The proposed amendment to the

Technical Specifications (TSs) would make administrative changes to TS

5.2 and 5.5. These changes reflect organizational changes in OPPD

senior management, delete specific titles of personnel on the Safety

Audit and Review Committee (SARC) and Plant Review Committee (PRC), and

make changes to SARC reviews and audits.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are administrative changes to reflect

organizational changes in Omaha Public Power District (OPPD) Senior

Management, remove specific titles from the membership of the Plant

Review Committee (PRC) and the Safety Audit and Review Committee

(SARC), add minor clarifications to SARC reviews and audits and

delete statements concerning the frequency of SARC audits from the

Technical Specifications (TS).

The proposed change to revise the overall corporate

responsibility for plant nuclear safety from the Senior Vice

President to Vice President is administrative in nature as it only

reflects an organizational change. Section 12 of the Updated Safety

Analysis Report describes the management structure and reporting

responsibilities of OPPD. Section 12 provides an organizational

chart to differentiate the Vice President in charge of nuclear

activities from other Vice Presidents within OPPD. Therefore,

changing the corporate reporting responsibility does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes to the membership of the PRC and SARC are

administrative in nature since only the specific titles of the

members are being removed from the TS. The management level and

expertise of personnel who are PRC or SARC members is not being

changed. The review of plant operations is still required to be in

compliance with ANSI N18.7-1976 and Regulatory Guide 1.33, Revision

2, as committed to in the Fort Calhoun Station Quality Assurance

(QA) Program. Any changes in the QA Program which reduce the

effectiveness of the program must be approved by the NRC in

accordance with 10 CFR 50.54(a)(3). Therefore, the proposed changes

to the membership of the PRC and SARC do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Clarifications of SARC reviews and audits and the deletion of

SARC audit frequencies from the TS are administrative changes. The

audit frequencies are required by the NRC approved QA Program and

any changes that could reduce the effectiveness of the QA Program

must be approved by the NRC in accordance with 10 CFR 50.54(a)(3).

Therefore, the clarifications and deletion of the specific audit

frequencies do not involve a significant increase in the probability

or consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of

accident from any accident previously evaluated.

The proposed changes are administrative in nature to reflect

organizational changes in OPPD Senior Management, remove specific

titles from the membership of the PRC and SARC, provide minor

clarifications of SARC reviews and audits and delete statements

concerning the frequency of SARC audits from the TS. The proposed

changes do not revise any equipment setpoints, change the manner in

which any plant equipment is operated, or propose any new operating

modes. Therefore, the proposed changes do not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

(3) Involve a significant reduction in a margin of safety.

The proposed changes revise organizational and administrative

requirements contained within the Administrative Controls section of

the TS. The proposed changes do not revise any equipment setpoints,

change the manner in which any plant equipment is operated, or

propose any new operating modes. Therefore, the proposed changes do

not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875

Connecticut Avenue, NW., Washington, DC 20009-5728

NRC Project Director: Theodore R. Quay

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: October 28, 1994

Description of amendment request: The proposed changes to the

Technical Specifications (TS) for the two units would add reference

20 (Unit 1) and reference 18 (Unit 2) to Section

6.9.3.2 as ``PL-NF-90-001, Supplement 1, 'Application of Reactor

Analysis Methods for BWR Design and Analysis: Loss of Feedwater Heating

Changes and Use of RETRAN MOD 5.1', September 1994''. These changes

would add changes to the methodology that the licensee is using to

perform its nuclear fuel reload analysis for the two units.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

I. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Incorporation of these proposed minor changes into PP&L's NRC

approved methodology for performing reload licensing analysis is

considered to be an enhancement to the currently approved

methodology. Upgrading of the RETRAN code allows for taking

advantage of state-of-the-art technology, while utilization of the

generic correlation for the LOFWH event supports consistency in

licensing analysis performance. Results of incorporating these

changes will not significantly increase the probability or the

consequences of an accident previously evaluated.

II. Create the possibility of a new or different kind of

accident from any accident previously evaluated.

As stated above, the incorporation of these minor changes are

considered enhancements, allowing PP&L to more efficiently and cost

effectively continue to perform future reload licensing analysis.

Therefore, the incorporation of these changes will not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

III. Involve a significant reduction in a margin of safety.

In addition to the extensive testing perform by EPRI, PP&L has

performed its own comparison tests utilizing RETRAN MOD005.1 in

place of MOD004 for four licensing transients that use the RETRAN

code. Results of this comparison were essentially the same for both

codes and support this proposed change. Also, the Loss of Feedwater

Heating event is not a limiting event for establishing MCPR

Operating Limits for Susquehanna. Therefore, the incorporation of

these changes will have no impact on current safety margins, nor

will they involve a significant reduction in the margin to safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

and Trowbridge, 2300 N Street NW., Washington, DC 20037

NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: October 28, 1994

Description of amendment request: The proposed changes to the

Technical Specifications (TS) for the two units would make a number of

administrative changes. These would include changing the title of the

positions of Superintendent of Plant to the Vice President-Nuclear

Operations, and changing the title of Vice-President-Nuclear Operations

to Senior Vice President-Nuclear for the listing of the assignment of

certain duties in various sub-sections of Section 6.0 of the TS. Other

proposed changes would be the deletion of a number of footnotes

indicating times, dates, and events that are no longer applicable, the

addition of a footnote to Section 6.5.1.2 indicating that the Station

Duty Manager shall act as a PORC [Plant Operations Review Committee]

chairman in the absence of the Vice President-Nuclear Operations, and

the change of the Semiannual Radioactive Effluent Release Report to

Annual Radioactive Effluent Release Report in Table 4.11.2.1.2-1

footnote g.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposal to change the title of Superintendent of Plant to

Vice President-Nuclear Operations and Vice President-Nuclear

Operations to Senior Vice President-Nuclear (for certain duties) is

administrative in nature and does not compromise the minimum

qualifications or training required for these positions. Therefore,

the proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change for the removal of footnotes that reference

periods of time, dates, and events that have since past is justified

based on the fact that they are no longer applicable. Because

operators must perform unnecessary applicability reviews on these no

longer applicable footnotes, removing the footnotes decreases the

potential for confusion and incorrect actions. Therefore, the

proposed change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed change to add a footnote indicating the Station

Duty Manager shall act as PORC chairman in the absence of the Vice

President-Nuclear Operations will ensure continuous leadership of

PORC and will enhance performance by ensuring a responsible

individual is available during all shifts. Therefore, the proposed

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

Changing the Semiannual Radioactive Effluent Release Report to

Annual Radioactive Effluent Release Report was previously approved

in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating

this change into footnote g of Table 4.11.2.1.2-1 will maintain

accuracy and consistency. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposal to change the title of Superintendent of Plant to

Vice President-Nuclear Operations and Vice President-Nuclear

Operations to Senior Vice President-Nuclear (for certain duties) is

administrative in nature and does not compromise the minimum

qualifications or training required for these positions. Also, the

change does not diminish the responsibilities or functions of these

positions. Therefore, the proposed change does not create the

possibility of a new or different kind of accident from any accident

previously evaluated.

The proposed change for the removal of footnotes that reference

periods of time, dates, and events that have since past is justified

based on the fact that they are no longer applicable. Because

operators must perform unnecessary applicability reviews on them,

removing the no longer applicable footnotes decreases the potential

for confusion and incorrect actions, thereby enhancing the safe

operation of Susquehanna SES. Therefore, the proposed change does

not create the possibility of a new or different kind of accident

from any accident previously evaluated.

The proposed change to add a footnote indicating the Station

Duty Manager shall act as PORC chairman in the absence of the Vice

President-Nuclear Operations will ensure continuous leadership of

PORC and will enhance performance by ensuring a responsible

individual is available during all shifts. Therefore, the proposed

change does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

Changing Semiannual Radioactive Effluent Release Report to

Annual Radioactive Effluent Release Report was previously approved

in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating

this change into footnote g of Table 4.11.2.1.2-1 will maintain

accuracy and consistency. Therefore, the proposed change does not

create the possibility of a new or different kind of accident from

any accident previously evaluated.

III. This change does not involve a significant reduction in a

margin of safety.

For the reasons discussed in items I and II above, as well as

the enclosed Safety Assessment, the proposed change does not involve

a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

and Trowbridge, 2300 N Street NW., Washington, DC 20037

NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: October 28, 1994

Description of amendment request: The amendment would delete the

requirements for chlorine detection and the associated Bases from the

Technical Specifications for each unit as a result of the removal of

bulk quantities of gaseous chlorine from the Susquehanna Steam Electric

Station. Specifically, Sections 3.3.7.8 and the associated Surveillance

Requirements in Section 4.3.7.8 would be deleted. In addition, Bases 3/

4.3.7.8 would also be deleted.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

Review of the various design basis accidents identified in

Chapter 15 of the Susquehanna SES Final Safety Analysis Report

(FSAR) concluded that none of these accidents are affected by

deletion of the chlorine detection requirements from Technical

Specifications. With the elimination of bulk quantities of gaseous

chlorine from use at Susquehanna SES the probability of control room

inhabitability due to a gaseous chlorine release has actually

decreased. Therefore, this proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed change involves only the deletion of the chlorine

detection system Technical Specifications based upon a plant

modification to remove gaseous chlorine as a biocide from

Susquehanna SES and replace it with a nonoxidizing biocide. The

release of chlorine from an off-site source is bounded by Reg. Guide

1.95 in that manual isolation capability for the control room

ventilation system is acceptable. Therefore, the proposed change

does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

III. This change does not involve a significant reduction in a

margin of safety.

The proposed change would not alter the margins of safety

provided in the existing FSAR analysis (Sections 2.2.3.1.3 and 6.4)

for chlorine release events since the basis for the existing margin

of safety, which are the Reg. Guide 1.95 requirements, are not

altered by the change. As stated above, since gaseous chlorine is no

longer used for open cooling water treatment at Susquehanna SES and

since the nonoxidizing biocide is relatively nontoxic to humans,

safety margin has actually increased. Therefore, the proposed change

does not involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

and Trowbridge, 2300 N Street NW., Washington, DC 20037

NRC Project Director: John F. Stolz

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: November 11, 1994

Description of amendment request: The amendment would extend the

Main Turbine Valve surveillance test interval from a weekly basis to no

greater than 92 days for all Main Turbine Stop, Control, and Combined

Intermediate Valves.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

I. This proposal does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed Technical Specification change to a quarterly

turbine inlet valve surveillance test interval is based on

maintaining the turbine missile generation probability within the

NRC criteria as stated in Table 3.1 of NUREG-1048. This, combined

with the NRC acceptable strike-and-damage probability as specified

in NUREG-1048, will keep the probability of unacceptable damage to

safety-related structures, systems, and components from turbine

missiles acceptably low (i.e., -7) . Thus, the NRC

acceptable risk rate of -7/yr. is not changed and there is

no increase in the probability of an accident previously evaluated.

The proposed Technical Specification change to the turbine inlet

valve surveillance interval does not effect the sequence of events

or the consequences of an accident previously evaluated. The

surveillance interval does not affect the strike and damage scenario

of an accident previously evaluated. Thus, the radiological

consequences of an accident previously evaluated will not be

increased.

II. This proposal does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed Technical Specification change to the turbine inlet

valve surveillance interval does not affect the surveillance test

characteristics. There are no new surveillance testing requirements.

Surveillance testing of these valves does not create the possibility

for a new or different kind of accident from any accident previously

evaluated.

III. This change does not involve a significant reduction in a

margin of safety.

The proposed Technical Specification change to the turbine inlet

valve surveillance interval is based on maintaining the same margin

of safety as previously determined by the NRC and does not reduce

the margin of safety. In fact, the reduction in the testing rate

will reduce the potential for testing related transients, which have

been credited with causing 18 reactor scrams in the period 1985

through 1992.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre,

Pennsylvania 18701

Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts

and Trowbridge, 2300 N Street NW., Washington, DC 20037

NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public Service Electric and Gas

Company, Delmarva Power and Light Company, and Atlantic City

Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom

Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: November 14, 1994

Description of amendment request: The proposed changes relocate

audit topics and frequencies, Nuclear Review Board review requirements

and requirements associated with the independent Safety Engineering

Group function from the Peach Bottom Atomic Power Station, Units 2 and

3 Technical Specifications to licensee controlled documents.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:1)

The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the changes relocate requirements from the TS to licensee

controlled documents consistent with the NRC Final Policy Statement

on TS Improvements. Any changes to the licensee controlled documents

will be evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59

as appropriate. Therefore, these changes will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2) The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously evaluated

because the changes will not alter the plant or the manner in which

the plant is operated. The changes will not involve a design change

or introduce any new failure modes. The changes will not alter

assumptions made in the safety analysis and licensing basis.

Adequate control of information will be maintained. Therefore, these

changes will not create the possibility of a new or different kind

of accident from any accident previously evaluated.

2) The proposed changes do not involve a significant reduction

in a margin of safety because they have no impact on any safety

analysis assumptions. The requirements to be transposed from the TS

to licensee controlled documents are the same as the existing TS.

Any future changes to licensee controlled documents will be

evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59 as

appropriate. Because the proposed changes are consistent with NUREG-

1433, as modified by approved generic change BWOG-09, and the change

controls for proposed relocated details and requirements provide an

equivalent level of regulatory authority, revising the TS to reflect

the approved level of detail and requirements ensures no significant

reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

Pennsylvania 17105.

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, Philadelphia Electric Company, 2301 Market Street,

Philadelphia, Pennsylvania 19101

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

BrownsFerry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: November 15, 1994 (TS 350)

Description of amendment request: The proposed change would remove

the frequency for each of the audits specified in the administrative

controls section of the technical specifications (TS). The requirements

to perform the audits would be retained, but the frequency for their

performance would be controlled by a requirement to be added to the

Nuclear Quality Assurance Plan. This would require that the audits

listed in the TS be performed on a biennial frequency. In addition, the

proposed change would remove the requirement to perform site

Radiological Emergency Plan and Physical Security/Safeguard Contingency

Plan reviews and audits from the TS, since these requirements presently

exist in the respective Plans.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

TVA has concluded that operation of BFN units 1, 2, and 3 in

accordance with the proposed change to the technical specifications

does not involve a significant hazards consideration. TVA's

conclusion is based on its evaluation in accordance with 10 CFR

50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).

TVA's conclusion is based on the following:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The likelihood that an accident will occur is neither increased

or decreased by this Technical Specification change which only

affects review and audit frequencies. This Technical Specification

change will not impact the function or method of operation of plant

equipment. Thus, there is not a significant increase in the

probability of a previously analyzed accident due to this change. No

systems, equipment, or components are affected by the proposed

change. Thus, the consequences of a malfunction of equipment

important to safety previously evaluated in the UFSAR are not

increased by this change.

The proposed change only affects review and audit frequencies.

As such, the proposed change has no impact on accident initiators or

plant equipment, and thus, does not affect the probabilities or

consequences of an accident.

Therefore, we conclude that this change does not significantly

increase the probabilities or consequences of an accident.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed changes do not involve changes to the physical

plant or operations. Since program audits do not contribute to

accident initiation, a change related to audit functions cannot

produce a new accident scenario or produce a new type of equipment

malfunction. Also, this change does not alter any existing accident

scenarios. The proposed change does not affect equipment or its

operation, and, thus, does not create the possibility of a new or

different kind of accident. Therefore, the proposed change does not

create the possibility of a new or different kind of accident.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change concerning conduct of reviews and audits

does not directly affect plant equipment or operation. Safety limits

and limiting safety system settings are no affected by this proposed

change.

Therefore, use of the proposed Technical Specification would not

involve any reduction in the margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of amendment request: November 15, 1994 (TS 94-12)

Description of amendment request: The proposed change would remove

the frequency for each of the audits specified in the administrative

controls section of the technical specifications (TS). The requirements

to perform the audits would be retained, but the frequency for their

performance would be controlled by a requirement to be added to the

Nuclear Quality Assurance Plan. This would require that the audits

listed in the TS be performed on a biennial frequency. In addition, the

proposed change would remove the requirement to perform site

Radiological Emergency Plan and Physical Security/Safeguard Contingency

Plan reviews and audits from the TS, since these requirements presently

exist in the respective Plans.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The standards used to arrive at a determination that a Technical

Specification change request involves no significant hazards

consideration are included in the Commission's regulations, 10 CFR

50.92, which states that no significant hazards considerations are

involved if the operation of the facility in accordance with the

proposed amendment would not: (1) involve a significant increase in

the probability or consequences of an accident previously evaluated;

or (2) create the possibility of a new or different kind of accident

from any accident previously evaluated; or (3) involve a significant

reduction in a margin of safety. Each standard is addressed as

follows:

1. Operation of the facility in accordance with the proposed

technical specifications would not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The likelihood that an accident will occur is neither increased

or decreased by the Technical Specification change which only

affects review and audit frequencies. This Technical Specification

change will not impact the function or method of operation of plant

equipment. Thus, there is not a significant increase in the

probability of a previously analyzed accident due to this change. No

systems, equipment, or components are affected by the proposed

changes. Thus, the consequences of a malfunction of equipment

important to safety previously evaluated in the FSAR are not

increased by this change.

The proposed change only affects review and audit frequencies.

As such, the proposed change has no impact on accident initiators or

plant equipment, and thus, does not affect the probabilities or

consequences of an accident.

Therefore, we conclude that this change does not significantly

increase the probabilities or consequences of an accident.

2. Operation of the facility in accordance with the proposed

technical specifications would not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes do not involve changes to the physical

plant or operations. Since program audits do not contribute to

accident initiation, a change related to audit functions cannot

produce a new accident scenario or produce a new type of equipment

malfunction. Also, this change does not alter any existing accident

scenarios. The proposed change does not affect equipment or its

operation, and, thus, does not create the possibility of a new or

different kind of accident. Therefore, the proposed change does not

create the possibility of a new or different kind of accident.

3. Operation of the facility in accordance with the proposed

technical specifications would not involve a significant reduction

in a margin of safety.

The proposed change concerning conduct of reviews and audits

does not directly affect plant equipment or operation. Safety limits

and limiting safety system settings are no affected by this proposed

change.

Therefore, use of the proposed Technical Specification would not

involve any reduction in the margin of safety.

The NRC has reviewed the licensee's analysis and, based on

thisreview, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: September 8, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification 4.2.2.2, 4.2.2.4, and 6.9.19 to

incorporate a penalty in the Core Operating Limit Report (COLR) to

account for FQ increases greater than 2 percent between

measurements.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes to the Technical Specifications do not

involve a significant hazards consideration because operation of

Callaway Plant in accordance with these changes would not:

1) Involve a significant increase in the probability of

occurrence or the consequences of an accident or malfunction of

equipment important to safety previously evaluated in the safety

analysis report. There is no increase in the probability of

occurrence or the consequences of an accident. The removal of

FQ(Z) penalty values from the Callaway Plant Technical

Specifications and the creation of cycle-specific FQ(Z) values

in the COLR has no influence or impact on the probability or

consequences of any accident previously evaluated. The cycle-

specific FQ(Z) values, although not in Technical

Specifications, will be followed in the operation of the Callaway

Plant. The proposed amendment still requires exactly the same

actions to be taken when or if FQ(Z) limits are exceeded as is

required by current Technical Specifications.

2) Create a possibility of a new or different kind of accident

from any previously evaluated in the safety analysis report. There

is no new type of accident or malfunction created and the method and

manner of plant operation will not change. As stated earlier, the

removal of the cycle-specific FQ(Z) value has no influence or

impact, nor does it contribute in any way to the probability or

consequences of an accident. No safety-related equipment, safety

function, or plant operation will be altered as a result of this

proposed change. The cycle-specific FQ(Z) values are calculated

using NRC approved methods. The Technical Specifications will

continue to require operation within the required FQ(Z) limits

and appropriate actions will be taken when or if limits are

exceeded.

3) Involve a significant reduction in a margin of safety. This

is based on the fact that no plant design changes are involved and

the method and manner of plant operation remains the same. The

margin of safety is not affected by change and removal of FQ(Z)

penalty values from the Technical Specifications. The margin of

safety presently provided by current Technical Specifications

remains unchanged. The current FQ(Z) limits remain unchanged

and the current safety analysis limits remain valid and unaffected

by this change. The proposed amendment continues to require

operation within the core limits as obtained from the NRC approved

design methodology and appropriate actions to be taken when or if

FQ(Z) limits are violated remain unchanged.

Given the above discussions as well as those presented in the

Safety Evaluation, the proposed change does not adversely affect or

endanger the health or safety of the general public or involve a

significant safety hazard.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037

NRC Project Director: Leif J. Norrholm

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of amendment request: September 12, 1994

Description of amendment request: The proposed amendment would

revise Technical Specification 3.7.1.1 (Tables 3.7-1 and 3.7-2). Tables

2.2-1 and 3.3-2, and Bases 3/4.7. Tables 3.7-1, 3.7-2, 3.3-2, and 2.2-1

would be revised to provide appropriate margin to relax main steam line

safety valve setpoint tolerance. Bases 3/4.7 would be revised to

incorporate the methodology used to determine the maximum allowable

power level associated with inoperable main steam line safety valves.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed changes to the Technical Specifications do not

involve a significant hazards consideration because operation of

Callaway Plant in accordance with these changes would not:

1. Involve a significant increase in the probability of

occurrence or the consequences of an accident or malfunction of

equipment important to safety previously evaluated in the safety

analysis report. The main steam line safety valves are designed to

mitigate transients by preventing overpressurization of the main

steam system. The proposed change does not alter this design basis.

The revised analysis shows that the probability or

consequences of all previously analyzed accidents are not

changed by increasing the setpoint tolerance of the safety valves.

Therefore, there is no increase in the probability of occurrence or

the consequences of any accident.

2. Create the possibility of a new or different kind of accident

from any previously evaluated in the safety analysis report. There

is no new type of accident or malfunction created, the method and

manner of plant operation will not change nor is there a change in

the method in which any safety related system performs its function.

Any main steam safety valve lifting at the extremes of the proposed

tolerance will not result in low lift setpoint that is less than the

normal no load system pressure or a high lift setpoint that allows

main steam system overpressurization.

3. Involve a significant reduction in a margin of safety. This

is based on the fact that no plant design changes are involved and

the method and manner of plant operation remains the same. With the

increased setpoint tolerance, the main steam safety valves will

still prevent pressure from exceeding 110 percent of design pressure

in accordance with the ASME code. All FSAR accident analysis

conclusions remain valid and unaffected by this change.

Given the above discussions as well as those presented in the

Safety Evaluation, the proposed change does not adversely affect or

endanger the health or safety of the general public or involve a

significant safety hazard.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Callaway County Public

Library, 710 Court Street, Fulton, Missouri 65251.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

& Trowbridge, 2300 N Street, N.W., Washington, DC 20037

NRC Project Director: Leif J. Norrholm

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: July 9, 1993, with supplemental

information provided October 8, 1993, October 25, 1993, January 6,

1994, February 2, 1994, May 3, 1994, May 13, 1994, September 26, 1994,

and October 12, 1994.

Description of amendment request: The proposed amendment would

modify the operating license and several Technical Specifications (TS)

to allow an increase in licensed power level by 4.9%. This would raise

the licensed power level from the current 3323 MWt to 3486 MWt, but

will not affect the basic fuel design or fuel operating limits. The

uprate in power will be accomplished by expanding the existing power-

to-flow map to allow an increase in core flow along the flow control

lines, with an associated increase in core power. The increased flow

and power will also cause an increase in operating reactor vessel steam

dome pressure that will require an associated TS change.

Other TS changes proposed to reflect necessary modifications to

address the power uprate are (1) an increase in the average power range

monitor (APRM) and flow biased scram and rod block monitor (RBM)

setpoints, (2) an increase in reactor steam pressure limits, (3) an

increase in main steam line (MSL) isolation valve and tunnel high

differential temperature to reflect the increased operating pressure,

(4) revised temperature/pressure limit curves to reflect the higher

neutron flux over vessel life, (5) an increase in calculated peak

containment pressure, (6) an increase in the pressure at which reactor

core isolation cooling (RCIC) testing occurs, and (7) an increase in

the safety relief valve (SRV) setpoints.

The licensee, in the above reference letters, also proposed changes

to the TS that are not associated with the power uprate. These proposed

changes are (1) a change of the reactor protection system and End-of-

Cycle/Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a

variable setpoint based on power level, (2) an increase in the SRV

setpoint tolerance, and (3) an increase in the number of automatic

depressurization system (ADS) valves that are allowed out of service.

Although not part of the change to the license or associated TS,

the licensee is also revising the bases to reflect the TS changes and

power uprate.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92 (c). The NRC staff's review is

presented below:

1. Does the amendment involve a significant increase in the

probability or consequences of an accident previously evaluated?

* [Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]Plant

operation at a higher power level is accomplished by increasing reactor

core flow along flow control lines to achieve the desired increase in

steam flow to the turbine/generator. The maximum allowable reactor

recirculation flow rate remains unchanged from the original plant

design analysis. The increased reactor core flow requires a

corresponding increase in feedwater flow that remains well within the

design of the feedwater system components. The increased power also

requires a small increase in reactor pressure. Safety relief valve and

power, pressure, and flow-related instrumentation trip setpoints are

increased slightly to accommodate the power uprate, to maintain

approximately the same level of trip avoidance and safety system

challenges as before the uprated power condition.

The plant is operated in the same manner at uprated power as it is

at the currently licensed power level, including the methods and

sequences of system and component operation. Since the level of trip

avoidance and safety system challenges remains approximately the same,

the frequency of operational responses to these events is not

increased. Reactor fuel operating limits, designed to protect the fuel

cladding, are maintained and thus provide the same level of protection

as before the uprated power condition. The original design and

regulatory criteria established for plant equipment, including ASME

code, IEEE standards, NEMA standards, and Regulatory Guide criteria,

are still imposed and met for operation at the uprated power level. In

addition, the reactor vessel and internals, reactor connecting piping,

balance of plant piping, primary containment, and related systems and

components still meet the pre-uprate design and licensing criteria. The

power uprate does not change the likelihood of failure of these systems

or components. Thus, the probability of an accident previously

evaluated is not significantly affected by the proposed power uprate.

The consequences of postulated, power-dependent accidents are

proportional to the power level assumed in the safety analysis. This is

because potential offsite doses increase proportionately to reactor

power since the radiological source term is directly proportional to

reactor power. The meteorological factors are unaffected by the

proposed power uprate. The current accident analyses are based on

104.4% of original rated power. The accident analyses for power uprate

are based on 1.02 x 104.9% (or 107%) of original rated power. Thus,

power uprate increases postulated consequences of an accident by less

than 3% over previous postulated consequences, while still remaining

well within the 10 CFR Part 100 limits. In addition, a spectrum of

hypothetical accidents and transients were investigated for power

uprate, and the bounding events met the same regulatory criteria to

which WNP-2 is currently licensed, including Maximum Average Planar

Linear Heat Generation Rate (MAPLHGR), Operating Limit Minimum Critical

Power Ratio (OLMCPR), 10 CFR 50.46 and 10 CFR Part 50-Appendix K, for

fuel cladding integrity, and containment criteria in 10 CFR Part 50-

Appendix A, Criterion 38 and Criterion 50.

The results of these analyses demonstrate that operation at the

proposed power uprate level does not significantly increase the

probability or consequences of any accident previously evaluated.

* [Increase in the maximum allowable reactor steam dome operating

pressure limit]

The operating pressure limit is increased by the same amount as the

nominal operating pressure increase for power uprate. This change to

the dome operating pressure limit is consistent with and meets the

current design criteria used for evaluation of steady state operating

conditions and for the most limiting event for vessel overpressure

protection. Operation at a higher pressure results in the plant

operating closer to the criteria used in the design analysis. However,

since operation of the plant within design limits is considered to

result in a very low probability of failure of systems or components,

this small increase in maximum operating pressure is considered to have

a negligible increase in failure probability. Thus the proposed change

does not significantly increase the likelihood of failure of existing

systems or components, and does not significantly affect the

probability of an accident previously evaluated.

The power uprate overpressure protection analysis results show the

peak reactor pressure vessel (RPV) pressure will remain below the ASME

code limit, keeping any postulated radiological consequences within the

bounds of existing analyses. Thus the consequences of an accident

previously evaluated are not significantly affected.

* [Increase in the average power range monitor (APRM) and flow

biased scram and rod block monitor (RBM) setpoints]

These scrams and rod blocks are designed to prevent fuel damage due

to power during postulated events or anticipated operational

occurrences. The setpoints for these scrams and rod blocks are

increased by the same amount of the proposed increase in licensed

power. The change in these setpoints does not affect the operation of

any system or component, nor does it affect the circuitry that provides

the protective function. Thus the increase in setpoints does not affect

the probability of any accident previously evaluated.

The increased setpoints maintain the same difference between

licensed power level and scram setpoints, and between the scram line

and rod block lines on the extended load line limit curve. The

increased setpoints result in higher postulated source terms for

accidents previously analyzed. The proposed change does not affect the

response or operation of mitigative equipment. The licensee's analyses

supporting power uprate demonstrate that the increased values do not

significantly affect the consequences of an accident. The proposed

change also maintains the current level of scram avoidance with

associated avoidance of unnecessary challenges to plant equipment,

while providing protection for the fuel, reactor systems, and

containment to meet current design requirements. Thus the proposed

change does not significantly increase the consequences of accident

previously analyzed.

* [Increase in main steam line (MSL) high flow differential

pressure setpoint to reflect the increased operating pressure]

The MSL high flow differential pressure setpoint is increased to

reflect the higher steam flows necessary to operate the plant at the

higher power. The proposed change does not affect the design,

construction, or operation of existing plant equipment, nor does it add

new equipment. The proposed change does not, therefore, affect the

probability of accidents previously analyzed. The increased setpoint

does not affect the maximum closure time for the main steam isolation

valves, thus the release of fission products during postulated accident

is not changed from current design analyses. The proposed change does

not, therefore, affect the consequences of accidents previously

analyzed.

* [Revised temperature/pressure limit curves to reflect the higher

neutron flux over vessel life]

The temperature/pressure limit curves are being modified to reflect

the increased exposure to neutron flux over the life of the vessel, to

retain the existing margin to brittle fracture over vessel life. The

plant will continue to be operated in conformance to the pressure/

temperature limits, retaining the existing probability of overcooling

events and associated postulated brittle fracture of the reactor

vessel. If a failure were to occur, the mode of failure is unaffected

by the proposed change, thus the consequences of a postulated failure

of the reactor vessel is unchanged by the proposed amendment.

* [Increase in the pressure at which reactor core isolation cooling

(RCIC) testing occurs]

The pressure at which the RCIC system is tested is being increased

to ensure adequate system operation with the increased reactor system

pressure required for power uprate. No credit for RCIC system operation

is taken in any accident analysis nor is RCIC included as an accident

initiator, thus this change does not affect the probability or

consequences of an accident previously evaluated.

* [Change of the reactor protection system and End-of-Cycle/

Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a

variable setpoint based on power level]

This setpoint is being changed from a specific value of main

turbine first stage pressure to a value to be calculated based on

variable plant conditions. The basis for the setpoint will remain

unchanged at a value equivalent to thermal power less than 30% of rated

thermal power. This change is proposed because turbine first stage

pressure can vary for a given thermal power level depending on the

amount of subcooling in the core, which itself is highly dependent on

feedwater temperatures. This proposed change does not affect the

reliability of operation of the trip circuitry, nor does it affect the

design or operation of plant equipment or safety systems. Thus the

change does not affect the probability of accidents previously

evaluated. In addition, the proposed change retains the same basis for

establishing the setpoint as currently stated in the TS, thus the

setpoint will be established at the level (30% rated thermal power)

assumed in the accident analysis. Thus the proposed change does not

affect the consequences of any accidents previously evaluated.

* [Increase in the safety relief valve (SRV) setpoints]

The licensee proposes to increase the setpoints of the two lowest-

set SRVs to accommodate the change in maximum operating pressure after

power uprate. This increase in SRV setpoints maintains approximately

the same difference between maximum operating RPV pressure and the

lowest SRV setpoint as currently exists. As discussed in the above

section regarding increased operating pressure, increasing the pressure

at which the lowest SRV actuates would allow operation at a slightly

higher pressure than currently allowed. This results in the plant

operating closer to the criteria used in the design analysis. However,

since operation of the plant within design limits is considered to

result in a very low probability of failure of systems or components,

this small increase in maximum operating pressure is considered to have

a negligible increase in failure probability. Thus the proposed change

does significantly increase the likelihood of failure of existing

systems or components, and does not affect the probability of an

accident previously evaluated.

With the increased SRV setpoint, the analysis results show that the

peak RPV pressure will remain below the ASME code limit, keeping any

postulated radiological consequences within the bounds of existing

analyses. Thus the consequences of an accident previously evaluated are

not affected.*

[Increase the value of Pa (the pressure at which primary

containment is tested). Add a new definition for Pa in the

``Definition'' section of the TS that gives a specific value of

Pa, and simplify the TS by deleting the specific value of Pa

and 1.10Pa from other locations in the TS]

The analysis to support the power uprate resulted in a higher

calculated peak containment pressure in response to postulated

accidents. To maintain the validity of the radiological analysis, the

containment leak rate testing must be based on a pressure greater than

or equal to the peak calculated containment pressure. The proposed

change does not affect the design, construction, or operation of

existing plant systems or components, and therefore does not affect the

probability of accidents previously evaluated.

The increase in test pressure does not affect the acceptance

criteria for the test, which is based on acceptable leakage. Offsite

dose projections are based in part on the leakage criteria, and since

the proposed change does not affect the leakage limits, the change does

not affect the consequences of accidents previously evaluated.

The addition of a definition of Pa, the delineation of the

specific value in the definition and associated deletion of the

specific value in individual TS, is a purely administrative change that

does not affect plant design, construction, or operation. This part of

the proposed change does not, therefore, affect the probability or

consequences of an accident previously evaluated.

* [Increase in the SRV setpoint tolerance]

The licensee is proposing to increase the setpoint tolerance for

the SRVs from +1/-3% to plus or minus 3% of the setpoint. This proposed

change does not affect how the SRVs operate in response to accidents or

abnormal operating occurrences, and thus does not affect the

probability of an accident previously evaluated.

The proposed increase in setpoint tolerance results in a

potentially higher pressure at which an SRV would lift. This increase

in pressure is approximately 23 psig, based on the proposed maximum

system pressure and proposed setpoint tolerance. This 2% increase in

maximum lift pressure would result in a proportional increase in

possible offsite dose due to a postulated event. The dose increase

would be something less than 2%, since the increased pressure would

result in a corresponding increase in differential pressure, with

attendant flow losses and subsequent filtration through accident

mitigation systems accounting for the lower dose. This small increase

in postulated offsite dose is not considered a significant increase in

the consequences of accidents previously evaluated.

* [Increase in the number of automatic depressurization system

(ADS) valves that are allowed out of service]

The proposed change would allow one ADS valve to be out of service

without time limit, and changing the minimum number of ADS valves

required to be in service from seven to six. The change would allow two

ADS valves (compared to the current one) to be out of service for up to

14 days, and would require plant shutdown within 12 hours for three or

more ADS valves (compared to the current two or more) out of service.

The proposed change would reduce the likelihood of an inadvertent

opening ADS/SRV as an initiating event if one SRV were out of service.

Having more than one ADS/SRV out of service would also reduce this

likelihood, although the reduction would have minimal effect since the

probability of multiple ADS/SRVs being out of service with a concurrent

lift and failure to close of an SRV is small. Thus the proposed change

would have little effect on the probability of accidents previously

evaluated.

The ADS/SRVs serve to limit overpressurization of the RPV and

connected primary piping. Having ADS valves out of service would

increase the consequences compared to the current TS which limit any

ADS valves out of service for more than 14 days. Reanalysis of the

effect of ADS valves on protection of the RPV demonstrated that five

ADS valves would prevent RPV overpressurization as the previous

analysis using six ADS valves. Thus the proposed change would not

affect the consequences of accidents previously evaluated.

2. Does the amendment create the possibility of a new or different

kind of accident from any accident previously evaluated?

* [Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]

Equipment that could be impacted by power uprate has been evaluated

by the licensee. The proposed change has not introduced any new

operating modes, equipment lineups, accident scenarios, or equipment

failure modes. The full spectrum of accident considerations defined in

Regulatory Guide 1.70 has been reviewed, and no new or different kind

of accident has been identified. Power uprate uses existing technology

and applies it within the capabilities of existing plant equipment in

accordance with existing regulatory criteria including NRC-approved

codes, standards, and methods. General Electric has designed to higher

power levels than the uprated power of WNP-2, and no new power-

dependent accidents have been identified. Therefore, the proposed

change does not create the possibility of a new or different kind of

accident from any accident previously evaluated.

* [Increase in the maximum allowable reactor steam dome operating

pressure limit]

The proposed change does not involve the addition of new equipment,

nor has it introduced any new operating modes, equipment lineups,

accident scenarios, or equipment failure modes. The proposed change

does not, therefore, create the possibility of a new or different kind

of accident from any accident previously evaluated.

* [Increase in the average power range monitor (APRM) and flow

biased scram and rod block monitor (RBM) setpoints]

The proposed setpoint changes do not involve the addition of new

equipment, nor do they introduce any new operating modes, equipment

lineups, accident scenarios, or equipment failure modes. The proposed

changes do not, therefore, create the possibility of a new or different

kind of accident from any accident previously evaluated.

* [Increase in main steam line (MSL) high flow differential

pressure setpoint to reflect the increased operating pressure]

The proposed setpoint change does not involve the addition of new

equipment, nor does it introduce any new operating modes, equipment

lineups, accident scenarios, or equipment failure modes. The proposed

change does not, therefore, create the possibility of a new or

different kind of accident from any accident previously evaluated.

* [Revised temperature/pressure limit curves to reflect the higher

neutron flux over vessel life]

The proposed change does not involve the addition of new equipment,

nor does it introduce any new operating modes, equipment lineups,

accident scenarios, or equipment failure modes. The proposed change

does not, therefore, create the possibility of a new or different kind

of accident from any accident previously evaluated.

* [Increase in the pressure at which reactor core isolation cooling

(RCIC) testing occurs]

The proposed change does not involve the addition of new equipment,

nor does it introduce any new operating modes, equipment lineups,

accident scenarios, or equipment failure modes. The proposed change

does not, therefore, create the possibility of a new or different kind

of accident from any accident previously evaluated.

* [Change of the reactor protection system and End-of-Cycle/

Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a

variable setpoint based on power level]

The proposed change does not involve the addition of new equipment,

nor does it introduce any new operating modes, equipment lineups,

accident scenarios, or equipment failure modes. The proposed change

does not, therefore, create the possibility of a new or different kind

of accident from any accident previously evaluated.

* [Increase in the safety relief valve (SRV) setpoints]

The proposed change does not involve the addition of new equipment,

nor does it introduce any new operating modes, equipment lineups,

accident scenarios, or equipment failure modes. The proposed change

does not, therefore, create the possibility of a new or different kind

of accident from any accident previously evaluated.

* [Increase the value of Pa (the pressure at which primary

containment is tested). Add a new definition for Pa in the

``Definition'' section of the TS that gives a specific value of

Pa, and simplify the TS by deleting the specific value of Pa

and 1.10 Pa from other locations in the TS]

The proposed change does not involve the addition of new equipment,

nor do it introduce any new operating modes, equipment lineups,

accident scenarios, or equipment failure modes. The proposed change

does not, therefore, create the possibility of a new or different kind

of accident from any accident previously evaluated.

* [Increase in the SRV setpoint tolerance]

The proposed change does not involve the addition of new equipment,

nor does it introduce any new operating modes, equipment lineups,

accident scenarios, or equipment failure modes. The proposed change

does not, therefore, create the possibility of a new or different kind

of accident from any accident previously evaluated.

* [Increase in the number of automatic depressurization system

(ADS) valves that are allowed out of service]

The proposed change does not involve the addition of new equipment,

nor does it introduce any new operating modes, accident sequences, or

equipment failure modes. The proposed change does allow a new equipment

lineup by allowing operation with one ADS valve out of service with no

limitation, which is not allowed by the current TS. In addition, two

ADS valves can be out of service for an extended period of time (14

days), with shutdown within 12 hours required only for three or more

valves inoperable compared to the current two valves. This allowed

equipment lineup does not alter the operation of the ADS valves or

their impact as potential event initiators as discussed in the FSAR.

The proposed change does not, therefore, create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. Does the amendment involve a significant reduction in a margin

of safety?*

[Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]

The plant was originally designed for operation at 105% rated steam

flow. The proposed change therefore does not affect the fuel design or

safety limits. The Maximum Axial Power Linear Heat Generation Rate

(MAPLHGR) limits remain the same. The Operating Limit Minimum Critical

Power Ratio (OLMCPR) is expected to increase by approximately 2%, which

will ensure that the margin to the Safety Limit MCPR is not affected.

The entire plant design has been reviewed to ensure that plant

equipment will perform properly and will still meet original design and

licensing criteria. The safety margins prescribed by the Code of

Federal Regulations have been maintained by meeting the appropriate

regulatory criteria. The margins provided by the application of the

ASME design acceptance criteria have been maintained, as well as other

margin-assuring acceptance criteria.

The emergency core cooling system-loss of coolant accident analysis

was conservatively performed based on two ADS valves out of service and

power level corresponding to 110% of original rated steam flow (plus 2%

power uncertainty factor). The analyzed results remain well below the

safety margin established at the 2200 deg.F peak centerline temperature

regulatory limit.

The overpressurization and containment analyses were repeated based

on 110% of original rated steam flow, resulting in a slightly higher

peak reactor vessel pressure. The increased pressure remains below the

acceptance criteria for the design basis, which the licensee identified

as below the ASME code limit and below the applicable TS limit.

From the containment analysis, the peak containment pressure

increases from 34.7 psig to 35.1 psig. This remains below the

containment design pressure of 45 psig.

The postulated radiological doses of design basis events, including

the bounding analysis involving loss of coolant accident, were

calculated based on the uprate power level. The results remain within

the design basis established by 10 CFR Part 100.

Based on these considerations, the proposed increase in licensed

power level does not significantly reduce any margins of safety.

* [Increase in the maximum allowable reactor steam dome operating

pressure limit]

The maximum pressure is increased by the same amount as the nominal

operating pressure increase for power uprate. The increased pressure

remains below the acceptance criteria for the design basis, which the

licensee identified as below the ASME code limit and below the

applicable TS limit. Thus the proposed TS change does not reduce the

margin to safety.

* [Increase in the average power range monitor (APRM) flow biased

scram and rod block monitor (RBM) setpoints]

The APRM flow biased scram setpoints were increase by the same

percentage of power as the uprated power. This maintains the same

margin to the trip setpoint while maintaining the same scram avoidance

as originally designed. The current margin (9%) between the scram line

and rod block line is maintained for power uprate. Thus the proposed TS

changes do not reduce any margin to safety.

* [Increase in main steam line (MSL) high flow isolation

differential pressure setpoint to reflect the increased operating

pressure]

The revised safety analysis retains the current analytic basis of

140% of rated steam flow to ensure the same level of scram avoidance is

maintained. This results in approximately a 10% increase in the

differential pressure required to trip the plant. This change does not,

however, affect the assumed closure times for the main steam isolation

valves in design analyses. The closure times determine the analyzed

release of fission products during postulated accidents. Since the

closure time is unaffected by the proposed change, the change does not

affect the margin of safety.

* [Revised temperature/pressure limit curves to reflect the higher

neutron flux over vessel life]

The proposed increase in licensed power also increases the neutron

fluence on the reactor pressure vessel over the license period (40

years). The analysis for the pressure/temperature limit curves was

updated to incorporate the increased fluence, and the revised curves

maintain the same margin to postulated brittle fracture of the reactor

vessel as the current TS curves. The proposed change does not,

therefore, change a margin of safety.

* [Increase in the pressure at which reactor core isolation cooling

(RCIC) testing occurs]

The proposed change would increase the test pressure for RCIC to

conform to the increased system pressure resulting from the power

uprate. Increasing the test pressure periodically verifies that RCIC

will operate at the increased pressure resulting from power uprate. In

addition, RCIC is not credited in safety analyses for assuring that

margins of safety are maintained. The proposed change, therefore, does

not affect any margin of safety.

* [Change of the reactor protection system and End-of-Cycle/

Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a

variable setpoint based on power level]

The safety analyses assume an EOC/RPT at 30% power. This proposed

change will allow the setpoint to vary relative to turbine first stage

pressure (the current TS setpoint), but will assure that the setpoint

is set based on a fixed 30% power. This assures that the EOC/RPT

maintains the existing margin of safety.

* [Increase in the safety relief valve (SRV) setpoints]

The two low set SRV setpoints are being increased by the same

amount of operating system pressure increase resulting from the power

uprate. The SRVs at the new setpoints will relieve pressure to ensure

the reactor coolant system remains below the acceptance criteria for

the design basis, which the licensee identified as below the ASME code

limit and below the applicable TS limit. Thus the proposed TS change

does not reduce the margin to safety.

* [Increase the value of Pa (the pressure at which primary

containment is tested). Add a new definition for Pa in the

``Definition'' section of the TS that gives a specific value of

Pa, and simplify the TS by deleting the specific value of Pa

and 1.10Pa from other locations in the TS]

The pressure at which primary containment is to be tested is being

increased to reflect the analyzed results of the power uprate. The

increase will assure that the periodic verification of containment

integrity is performed at a pressure that assures that postulated

radioactive release rates remain within analyzed assumptions. This

assures that existing margins of safety are maintained.

* [Increase in the SRV setpoint tolerance]

The proposed increase in the SRV setpoint tolerance from +1/-3% to

plus or minus 3% of the setpoint has the potential to increase the

actual pressure at which the SRV would lift. This increase in pressure

is approximately 23 psig, or 2% of the design lift pressure. This

pressure is still well within the RPV design pressure that establishes

the safety margin. The proposed change would not, therefore, affect a

margin of safety.

* [Increase in the number of automatic depressurization system

(ADS) valves that are allowed out of service]

The proposed change results in an increase in the calculated peak

centerline temperature (PCT) for postulated accidents, specifically,

the loss of coolant accident (LOCA). The margin of safety for this

parameter at WNP-2 is 2200 deg.F, which is also the regulatory limit.

Maintaining PCT less than 2200 deg.F ensures cladding integrity,

including the safety margin established by regulation. The proposed

change does not, therefore, affect a margin of safety.

Based on this review, it appears that the three standards of

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352

Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC 20005-3502

NRC Project Director: Theodore R. Quay

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of amendment request: October 31, 1994

Description of amendment request: The proposed amendment would

modify the Technical Specifications (TS) to: (1) add two action

statements that would provide allowed outage times for either one or

both of the scram discharge volume (SDV) vent or drain valves less

stringent that the current requirements of TS 3.0.3.; and (2) change

the surveillance requirements for the SDV vent and drain valves to

conduct the testing during shutdown conditions rather than at power as

currently required.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

[Adding action statements for allowed outage times]

The primary functions of the SDV vent and drain valves are to

isolate the SDV following a scram to stop leakage of reactor coolant

past the CRD seals and to reopen following a scram to reset to drain

the reactor coolant from the SDV to the reactor building equipment

drain sump. The SDV is sized to accept CRD over piston discharge

water from all 185 CRDs. The SDV vent and drain valves reopen when

the scram signal is reset to provide assurance that there is

sufficient SDV volume available to accept the CRD discharge in

This text is long and has been trimmed here. Open the source document for the complete record.